Text

Er 15-019, Rev. 1, Cooper Nuclear Station Pressure and Temperature Limits Report (PTLR) for 32 Effective Full-Power Years (EFPY) (Non-Proprietary Version of Er 15-015)
	ML16084A183
Person / Time
Site:	Cooper
Issue date:	03/02/2016
From:	Barker T, Domikaitis S, Mcclure T; Entergy Corp
To:	; Office of Nuclear Reactor Regulation
Shared Package
ML16084A181	List: ML16084A182; ML16084A183; NLS2016014, Er 15-019, Supplement 1, BWRVIP-135, Revision 3: BWR Vessel and Internals Project, Integrated Surveillance Program (ISP) Data Source Book and Plant Evaluations;
References
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NLS2016014 Page 1 of74 Cooper Nuclear Station Pressure and Temperature Limits Report (PTLR) for 32 Effective Full-Power Years (EFPY) (Non-Proprietary version of ER 15-015)

Cooper Nuclear Station, Docket No. 50-298, License No. DPR-46

~

NUCLEAR QUALITY RELATED 3-EN-DC-147 I REV. 5C1

~=-Entergy MANAGEMENT INFORMATIONAL USE PAGE 1of73 ATTACHMENT 9.1 MANUAL Engineering Reports ENGINEERING REPORT COVER SHEET & INSTRUCTIONS Engineering Report No.15-019 Rev I

Page I

of 28 Engineering Report Cover Sheet Engineering Report

Title:

Cooper Nuclear Station Pressure and Temperature Limits Report (PTLR) for 32 Effective Full-Power Years (EFPY)

(Non-Proprietary version of ER 15-015)

Engineering Report Type: (3)

New

~

Revision D Cancelled D

Superseded D

Superseded by:

Rev I: Added Attachment 1 for Cooper specific pages from BWRVIP-135, Revision 3. Added reference to Cooper Nuclear Station in Title and updated References to include GL 96-03.

ECR No. NIA EC No. NIA (4) Report Origin:

~ CNS D Vendor Vendor Document No.: ________ _

(5) Quality-Related:

~ Yes D No Prepared by:

Tim McClure/

Date: 3-2.-l(p

/Sign)

Design Verified:

Stan Domikaitis/

Date: i-2-lt Design Verifie ame/Sign)

Reviewed by:

NIA Date:

Reviewer (Print Name/Sign)

Approved by: 10f s. Brorw ~2~'-

Date: 3/3//h Supervisor I Manager{Prit Name/Sign)

Section 1.0 2.0 3.0 4.0 5.0 6.0 Figure 1 Figure 2 Figure 3 Figure 4 Figure 5 Table 1 Table 2 Table 3 Table 4 Appendix A Supplement 1 Cooper Nuclear Station PTLR ER 15-019 Revision 1 Page 2 of28 Table of Contents Purpose 3

Applicability 3

Methodology 4

Operating Limits 5

Discussion 6

References 11 CNS P-T Curve A (Hydrostatic Pressure and Leak Tests) for 14 32EFPY CNS P-T Curve B (Normal Operation - Core Not Critical) for

15.

32EFPY CNS P-T Curve C (Normal Operation - Core Critical) for 16 32 EFPY Cooper Feedwater Nozzle Finite Element Model [14]

17 Cooper Core Differential Pressure Nozzle Finite Element Model [16]

18 CNS Pressure Test (Curve A) P-T Curves for 32 EFPY 19 CNS Core Not Critical (Curve B) P-T Curves for 32 EFPY 22 CNS Core Critical (Curve C) P-T Curves for 32 EFPY 25 CNS ART Calculations for 32 EFPY 26 Cooper Reactor Vessel Materials Surveillance Program 27 BWRVIP-135, Revision 3, Cooper Applicable Pages (45 pages) 28

1.0 Purpose Cooper Nuclear Station PTLR ER 15-019 Revision 1 Page 3 of28 The purpose of the Cooper Nuclear Station (CNS) Pressure and Temperature Limits Report (PTLR) is to present operating limits relating to:

1. Reactor Coolant System (RCS) Pressure versus Temperature limits during Heatup, Cooldown and Hydrostatic/Class 1 Leak Testing;

2. RCS Heatup and Cooldown rates;

3. RPV head flange boltup temperature limits.

This report has been prepared in accordance with the requirements of Licensing Topical Reports SIR-05-044, Revision 1-A contained within BWROG-TP-11-022-A, Revision 1 [1] and 0900876.401, Revision 0-A contained within BWROG-TP-11-023-A, Revision 0 [2].

It should also be noted that the P-T curves referenced in this PTLR have previously been approved by the NRC in Amendment 245 [24]. No changes are being made under this PTLR to the current P-T curves that were approved by the NRC [24] and currently in effect at CNS.

2.0 Applicability This report is applicable to the CNS RPV for up to 32 Effective Full-Power Years (EFPY).

The following CNS Technical Specification (TS) is affected by the information contained in this report:

TS RCS Pressure and Temperature (P-T) Limits TS Surveillance Requirements 3.0 Methodology The limits in this report were derived as follows:

Cooper Nuclear Station PTLR ER 15-019 Revision 1 Page4 of28

1. The methodology used is consistent with Reference [1] and Reference [2], which have been approved by the NRC in References [25] and [26], respectively.

2. The neutron fluence is calculated in accordance with NRC Regulatory Guide 1.190 (RG 1.190) [3], using the RAMA computer code, as documented in Reference [ 4].

3. The adjusted reference temperature (ART) values for the limiting beltline materials are calculated in accordance with NRC Regulatory Guide 1.99, Revision 2 [5], as documented in Reference [6].

4. The pressure and temperature limits were calculated consistent with Reference [l],

"Pressure-Temperature Limits Report Methodology for Boiling Water Reactors,", as documented in NPPD calculation NEDC 07-048, Reference [7].

5. This revision of the pressure and temperature limits is to incorporate the following changes:

Initial issue of PTLR revised to include BWRVIP-135 Cooper specific information.

Changes to the curves, limits, or parameters within this PTLR, based upon new irradiation fluence data of the RPV, or other plant design assumptions in the Updated Final Safety Analysis Report (UFSAR), can be made pursuant to I 0 CFR 50.59 [17], provided the above methodologies are utilized. The revised PTLR shall be submitted to the NRC upon issuance.

Changes to the curves, limits, or parameters within this PTLR, based upon new surveillance capsule data of the RPV or other plant design assumption modifications in the UFSAR, cannot be made without prior NRC approval. Such analysis and revisions shall be submitted to the NRC for review prior to incorporation into the PTLR.

4.0 Operating Limits Cooper Nuclear Station PTLR ER 15-019 Revision 1 Page 5 of28 The pressure-temperature (P-T) curves included in this report represent steam dome pressure versus minimum vessel metal temperature and incorporate the appropriate non-beltline limits and irradiation embrittlement effects in the beltline region.

The operating limits for pressure and temperature are required for three categories of operation:

(a) hydrostatic pressure tests and leak tests, referred to as Curve A; (b) core not critical operation, referred to as Curve B; and (c) core critical operation, referred to as Curve C.

Complete P-T curves were developed for 32 EFPY for Cooper Nuclear Station, as documented in Reference [7] and approved by the NRC in CNS Amendment 245 [24]. The CNS P-T curves for 32 EFPY are provided in Figures 1 through 3, and a tabulation of the curves is included in Tables 1 through 3. The adjusted reference temperature (ART) tables for the CNS vessel beltline materials are shown in Table 4 for 32 EFPY (Reference [6]). The resulting P-T curves are based on the geometry, design and materials information for the CNS vessel with the following conditions:

Heatup and Cooldown rate limit during Hydrostatic Class 1 Leak Testing at or near isothermal conditions (Figure 1: Curve A): ::; 25°F/hour1 [7].

Normal Operating Heatup and Cooldown rate limit (Figure 2: Curve B - non-nuclear heating, and Figure 3: Curve C -nuclear heating)::::; 100°F/hour2 [7].

RPV bottom head coolant temperature to RPV coolant temperature t1T limit during Recirculation Pump startUp: ~ 145°F.

Recirculation loop coolant temperature to RPV coolant temperature t1T limit during Recirculation Pump startup:~ 50°F.

RPV flange and adjacent shell temperature limit 2::'; 70°F [7].

1 Interpreted as the temperature change in any I-hour period is less than or equal to 25°F.

2 Interpreted as the temperature change in any I-hour period is less than or equal to 100°F.

Cooper Nuclear Station PTLR ER 15-019 Revision 1 Page 6 of28 To address the NRC condition regarding lowest service temperature in Reference [1] the minimum temperature is set to 70°F, which is equal to the RTNoT,rnax + 60°F, for all curves. [24]

5.0 Discussion The adjusted reference temperature (ART) of the limiting beltline material is used to adjust beltline P-T curves to account for irradiation effects. RG 1.99 [5] provides the methods for determining the ART. The RG 1.99 methods for determining the limiting material and adjusting the P-T curves using ART are discussed in this section.

The vessel beltline copper (Cu) and nickel (Ni) values were obtained from the evaluation of the CNS vessel plate, weld, and forging materials [6]; this evaluation included the results of two surveillance capsules for the representative plate material and three surveillance capsules for the representative weld material. The Cu and Ni values were used with Table 1 ofRG 1.99 to determine a chemistry factor (CF) per Paragraph 1.1 of RG 1.99 for welds. The Cu and Ni values were used with Table 2 of RG 1.99 to determine a CF per Paragraph 1.1 of RG 1.99 for plates and forgings. However, the fitted CF for the limiting plate (which is based on credible surveillance data) in the CNS vessel bounds the RG 1.99 CF. Therefore, the fitted CF is used for the limiting beltline plate.

The peak RPV ID fluence value of 1.41x10 18 n/cm2 at 32 EFPYused in the P-T curve evaluation were obtained from Reference [4] and are calculated in accordance with RG 1.190

[3]. These fluence values apply to the limiting beltline lower intermediate shell plate (Heat No.

C2307-2). The fluence values for the lower intermediate shell plate are based upon an attenuation factor of 0.72 for a postulated 1/4T flaw. As a result, the 1/4T fluence for 32 EFPY for the limiting lower intermediate shell plate is 1.02 x 1018 n/cm2 for CNS.

The P-T limits are developed to bound all ferritic materials in the RPV, including the consideration of stress levels from structural discontinuities such as nozzles. The water level instrument (WLI) nozzle is located in the lower-intermediate shell beltline plates [7]. The nozzle material is not ferritic, however the effect of the penetration on the adjacent shell is considered

Cooper Nuclear Station PTLR ER 15-019 Revision 1 Page 7 of28 according to the methodology in Reference [2]. The RPV ID fluence value of 2.94 x 1017 n/cm2 at 32 EFPY used in the P-T curve evaluation of the WLI nozzle was obtained from Reference [4]

and is calculated in accordance with RG 1.190 [3]. This fluence value applies to the limiting WLI nozzle (Heat No. EV-26067). The fluence value for the WLI nozzle is based upon an attenuation factor of 0. 72 for a postulated 1/4T flaw. As a result, the 1/4T fluence for 32 EFPY for the limiting WLI nozzle is 2.13 x 1017 n/cm2 for CNS. There are no additional forged or partial penetration nozzles in the extended beltline.

The P-T curves for the core not critical and core critical operating conditions at a given EFPY apply for both the 1/4T and 3/4T locations. When combining pressure and thermal stresses, it is usually necessary to evaluate stresses at the 1/4T location (inside surface flaw) and the 3/4T location (outside surface flaw). This is because the thermal graoient tensile stress ofinterest is in the inner wall during cooldown and is in the outer wall during heatup. However, as a conservative simplification, the thermal gradient stresses at the 1/4T location are assumed to be tensile for both heatup and cooldown. This results in the approach of applying the maximum tensile stresses at the 1/4T location. This approach is conservatiye because irradiation effects cause the allowable toughness at the 1/4T to be less than that at 3/4T for a given metal temperature. This approach causes no operational difficulties, since the BWR is at steam saturation conditions during normal operation, and for a given pressure, the coolant saturation temperature is well above the P-T curve limiting temperature. Consequently, the material toughness at a given pressure would exceed the allowable toughness.

For the core not critical curve (Curve B) and the core critical curve (Curve C), the P-T curves specify a coolant heatup and cool down temperature rate of :'.S 100°F /hour for which the curves are applicable. However, the core not critical and the core critical curves were also developed to bound Service Level A/B RPV thermal transients. For the hydrostatic pressure and leak test curve (Curve A), a coolant heatup and cooldown temperature rate of :'.S 25°F/hour must be maintained. The P-T limits and corresponding limits of either Curve A or B may be applied, if necessary, while achieving or recovering from test conditions. So, although Curve A applies

Cooper Nuclear Station PTLR ER 15-019 Revision 1 Page 8 of28 during pressure testing, the limits of Curve B may be conservatively used during pressure testing if the pressure test heatup/cooldown rate limits cannot be maintained.

The initial RT NOT, the chemistry (weight-percent copper and nickel), and ART at the 1/4T location for all RPV beltline materials significantly affected by fluence (i.e., fluence > 1017 n/cm2 for E > lMeV) are shown in Table 4 for 32 EFPY [6].

Per Reference [6] and in accordance with Appendix A of Reference [1], the CNS representative weld and plate surveillance materials data were reviewed from the Boiling Water Reactor Vessel and Internals Project (BWRVIP) Integrated Surveillance Program (ISP) [19]. The representative heat of the plate material (C2307-2) in the ISP is the same as the lower intermediate shell plate material in the vessel beltline region of CNS. For plate heat C2307-2, since the scatter in the fitted results is less than 1-sigma (17°F), the margin term (cr8 = 17°F) is cut in half for the plate material when calculating the ART. The representative heat of the weld material (20291) in the ISP is not the same as the limiting weld material in the vessel beltline region of CNS. Therefore, CFs from the tables in RGl.99 were used in the determination of the ART values for all CNS beltline materials except for plate heat C2307-2.

The only computer code used in the determination of the CNS P-T curves was the ANSYS finite element computer program:

ANSYS, Revision 5.3 [8] for the feedwater (FW) nozzle (non-beltline) pressure arid thermal down shock stresses.

Mechanical and PrepPost, Release 11.0 (Service Pack 1) [9] for the development of the generic WLI nozzle stress intensity factors in [2].

Mechanical APDL and PrepPost, Release 12.1 [10] for the FW nozzle (non-beltline)

thermal ramp stresses and the core differential pressure (DP) nozzle (bottom head) pressure stress distribution.

ANSYS finite element analyses were used to develop the stress distributions through the FW, WLI, and core DP nozzles, and these stress distributions were used in the determination of the

Cooper Nuclear Station PTLR ER 15-019 Revision 1 Page 9 of28 stress intensity factors for these nozzles [2, 13, 14, 16]. At the time that each of the analyses above was performed, the AN SYS program was controlled under the vendor's 10 CFR 50 Appendix B [ 11] Quality Assurance Program for nuclear quality-related work. Benchmarking consistent with NRC GL 83-11, Supplement I [12] was performed as a part of the computer program verification by comparing the solutions produced by the computer code to hand calculations for several problems.

The plant-specific CNS FW nozzle analysis was performed to determine through-wall pressure stress distributions and thermal stress distributions due to bounding thermal transients [13, 14].

Detailed information regarding the analysis can be found in References [13] and [14]. The following inputs were used as input to the finite element analysis:

With respect to operating conditions, stress distributions were developed for two bounding thermal transients. A thermal shock, which represents the maximum thermal shock for the FW nozzle during normal and upset operating conditions [13], and a thermal ramp were analyzed [14]. Potential leakage past the primary and secondary thermal sleeves is considered in the heat transfer calculations. The thermal down shock of 450°F, which is associated with the turbine roll transient during startup, produces the highest tensile stresses at the 1/4T location. Because operation is along the saturation curve, these stresses are scaled to reflect the worst-case step change due to the available temperature difference. It is recognized that at low temperatures, the available temperature difference is insignificant and could potentially result in a near zero stress distribution. Therefore, a minimum stress distribution is calculated based on the thermal ramp of 100°F/hour, which is associated with the shutdown transient. Therefore, the combination of the thermal down shock and thermal ramp stresses represent the bounding stresses in the FW nozzle associated with 100°F/hour heatup/cooldown limits associated with the P-T curves for the upper vessel FW nozzle region.

Heat transfer coefficients were given in the CNS FW nozzle design basis stress report and are a function of FW temperature and flow rate. Bounding, or larger, convection coefficients were used in the present P-T curve analysis [13, 14]. Therefore, the heat

Cooper Nuclear.Station PTLR ER 15-019 Revision I Page 10 of28 transfer coefficients used in the analysis bound the actual operating conditions in the FW nozzle at CNS.

A two-dimensional finite element model of the FW nozzle was constructed (Figure 4).

The pressure stresses are multiplied by a factor of2.5 to account for the 3-D effects [13].

Material properties were taken at 350°F, which is approximately the average temperature for the shutdown transient, from the 1989 ASME Code [15]. The use of temperature independent material properties is consistent with original design basis documents. Use of temperature dependent material properties is expected to have minimal impact on the results of the analysis.

The plant-specific CNS core DP nozzle analysis was performed to determine a through-wall.

pressure stress distribution [ 16]. Detailed information regarding the analysis can be found in Reference [ 16]. The following inputs were used as input to the finite element analysis:

No thermal transients were analyzed as part of the plant-specific core DP nozzle evaluation. Thermal stresses were addressed generically as specified in [1] with the use of a stress concentration factor of 3.0 to account for the discontinuity in the bottom head.

A two-dimensional finite element model of the core DP nozzle was constructed (Figure 5). Material properties were taken at 325°F from the vessel stress report [16]. The use of temperature independent material properties is consistent with original design basis documents.

Intial RTNoTvalues were reported in the ART calculation in amendment 120 [22].

6.0 References Cooper Nuclear Station PTLR ER 15-019 Revision 1 Page 11 of28

1. BWROG-TP-11-022-A, Revision 1, Pressure Temperature Limits Report Methodology for Boiling Water Reactors, August 2013.

2. BWROG-TP-11-023-A, Revision 0, Linear Elastic Fracture Mechanics Evaluation of General Electric Boiling Water Reactor Water Level Instrument Nozzles for Pressure -

Temperature Curve Evaluations, May_2013.

3. U. S. Nuclear Regulatory Commission, Regulatory Guide 1.190, "Calculational and Dosimetry Methods for Determining Pressure Vessel Neutron Fluence", March 2001.

4. Cooper Nuclear Station Calculation NEDC 07-032, Revision 3, "CNS Review of Trans Ware Calculations NPP-FLU-003-R-002, Revision 0, NPP-FLU-003-R-004, and NPP-FLU-003-R-005, Reactor Pressure Vessel Fluence Evaluation", April 2013 that incorporated TransWare Enterprises Report No. NPP-FLU-003-R-005,, Revision 0, "Non-Proprietary Version of Cooper Nuclear Station Reactor Pressure Vessel Fluence Evaluation," January 2011, SI File No. 1100445.201.

5. U. S. Nuclear Regulatory Commission, Regulatory Guide 1.99, Revision 2, "Radiation Embrittlement of Reactor Vessel materials", May 1988.

6. Cooper Nuclear Station Calculation NEDC07-045, Revision 2, "Review of SIA Calculation COOP-27Q-301, ~RTNoT and ART Evaluation", April 2013 that incorporated Structural Integrity Associates Calculation No. 1100445.301, Revision 1, "~RTNoT and ART Evaluation", July 2010.

7. Cooper Nuclear Station Calculation, NEDC07-048, Revision 6, "Revised Pressure Temperature Curves", April 2013 that incorporated Structural Integrity Associates Calculation No. 1100445.303, Revision 0, "Revised P-T Curve Calculation, August 2011.

8. ANSYS, Revision 5.3, ANSYS Inc., October 1996.

9. ANSYS Mechanical and PrepPost, Release 11.0 (w/ Service Pack 1), ANSYS, Inc., August 2007.

Cooper Nuclear Station PTLR ER 15-019 Revision 1 Page 12 of28

10. ANSYS Mechanical APDL and PrepPost, Release 12.1 x64, ANSYS, Inc., November 2009.

11. U. S. Code of Federal Regulations, Title 10,.Energy, Part 50, Appendix B, "Quality Assurance for Nuclear Power Plants and Fuel Reprocessing Plants".

12. U. S. Nuclear Regulatory Commission, Generic Letter 83-11, Supplement 1, "License Qualification for Performing Safety Analyses", June 24, 1999.

13.. Cooper Nuclear Station Calculation No. NEDC99-020, "Review of Structural Integrity Report SIR-99-069 and Calculations No. NPPD-13Q-301, NPPD-13Q-302, NPPD-13-Q-303, specifically Structural Integrity Associates Calculation No. NPPD-13Q-302, Revision 1, "Feedwater Nozzle Stress Analysis," June 1999.

14. Cooper Nuclear Station Calculation No. NEDC99-020, Structural Integrity Associates Calculation No. 1100445.302, Revision 0, "Finite Element Stress Analysis of Cooper RPV Feedwater Nozzle," June 2011.

15. ASME Boiler and Pressure Vessel Code,Section III, Division 1, Appendices, 1989 Edition.

16. Cooper Nuclear Station Calculation, NEDC07-048, Revision 6, "Revised Pressure Temperature Curves", April 2013 that incorporated Structural Integrity Associates Calculation No. 1100445.304, Revision 0, "Core Differential Pressure Nozzle Finite Element Model and Stress Analysis," August 2011.

17. U. S. Code of Federal Regulations, Title 10, Part 50, Section 59, "Changes, tests and experiments," Aug. 28. 2007.

18. U.S. Code of Federal Regulations, Title 10, Part 50, Appendix H, "Reactor Vessel Material Surveillance Program Requirements," Jan. 31, 2008.

19. BWRVIP-135, Revision 3: BWR Vessel and Internals Project, Integrated Surveillance Program (ISP) Data Source Book and Plant Evaluations. EPRI, Palo Alto, CA: 2014.

3002003144. SI File No. BWRVIP-135P. EPRI PROPRIETARY INFORMATION.

(See Supplement 1)

Cooper Nuclear Station PTLR ER 15-019 Revision 1 Page 13 of28

20. Letter NLS2002 l 04 dated December 31, 2002, "License Amendment Request to Adopt an Integrated Reactor Vessel Material Surveillance Program, Cooper Nuclear Station, NRC Docket No. 50-298, DPR-46", from M.T. Coyle (NPPD) to U.S. Nuclear Regulatory Commission, ADAMS Accession No. ML030080070, SI File No. 1400473.202.

21. BWRVIP-86, Revision 1-A: BWR Vessel and Internals Project, Updated BWR Integrated Surveillance Program (ISP) Implementation Plan. EPRI, Palo Alto, CA: 2012. 1025144.

EPRI PROPRIETARY INFORMATION.

22. Cooper Nuclear Station Amendment 120 as approved by the NRC on April 26, 1988.

(ML021360424)

23. Cooper Nuclear Station Amendment 201 as approved by the NRC on October 23, 2003.

(ML033090607)

24. Cooper Nuclear Station Amendment 245 as approved by the NRC on February 22, 2013.

(ML13032A526).

25. U.S. NRC Letter to BWROG, "Final Safety Evaluation for Boiling Water Reactor Owners' Group Topical Report BWROG-TP-11-022, Revision 1, November 2011, "Pressure-Temperature Limits Report Methodology for Boiling Water Reactors" (TAC NO. ME7649, MLl 3277 A557).

26. U.S. NRC Letter to BWROG, "Final Safety Evaluation for Boiling Water Reactor Owners' Group Topical Report BWROG-TP-11-023, Revision 0, November 2011, "Linear Elastic Fracture Mechanics Evaluation of General Electric Boiling Water reactor Water Level Instrument Nozzles for Pressure-Temperature Curve Evaluations" (TAC NO. ME7650, ML13183A017)

27. U. S. Nuclear Regulatory Commission, Generic Letter 96-03, "Relocation of the Pressure Temperature Limit Curves and Low Temperature Overpressure Protection System Limits",

January31, 1996.

Cooper Nuclear Station PTLR ER 15-019 Revision 1 Page 14 of28 Figure 1: CNS P-T Curve A (Hydrostatic Pressure and Leak Tests) for 32 EFPY [7]

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20 40 60 80 100 120 140 160 180 200 220 240 260 280 300 320 340 MINIMUM REACTOR VESSEL METAL TEMPERATURE {°F)

Cooper Nuclear Station PTLR ER 15-019 Revision 1 Page 15 of28 Figure 2: CNS P-T Curve B (Normal Operation-Core Not Critical) for 32 EFPY [7]

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20 40 60 80 100 120 140 160 180 200 220 240 260 280 300 320 340 MINIMUM REACTOR VESSEL METAL TEMPERATURE (°F)

Cooper Nuclear Station PTLR ER 15-019 Revision 1 Page 16 of28 Figure 3: CNS P-T Curve C (Normal Operation - Core Critical) for 32 EFPY [7]

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I ELE'MENI'S MAT NOM Cooper Nuclear Station PTLR ER 15-019 Revision 1 Page 17 of28 Figure 4: Cooper Feedwater Nozzle Finite Element Model (14)

J\\N APR 20 2011 15 :18 :48 PI.OT l\\D.

1

Cooper Nuclear Station PTLR ER 15-019 Revision 1 Page 18 of28 Figure 5: Cooper Core Differential Pressure Nozzle Finite Element Model [16) l J\\N 3

Co ope

Cooper Nuclear Station PTLR ER 15-019 Revision 1 Page 19 of28 Table 1: CNS Pressure Test (Curve A) P-T Curves for 32 EFPY [7]

Beltline Region P-T Curve P-T Curve Temperature Pressure 70.00 0

70.00 50 70.00 100 70.00 150 70.00 200 70.00 250 70.00 300 70.00 312 70.00 313 70.00 350 70.00 400 70.00 450 70.00 500 77.37 550 85.28 600 92.11 650 100.07 700 109.08 750 116.71 800 123.34 850 129.19 900 134.42 950 139.16 1000 143.49 1050 147.47 1100 151.16 1150 154.60 1200 157.82 1250 160.83 1300

Cooper Nuclear Station PTLR ER 15-019 Revision 1 Page 20of28 Table 1: CNS Pressure Test (Curve A) P-T Curves for 32 EFPY (continued)

Plant =.;;~::;,~::AN~)/',:\\

Component= ~ottoi)'l;HE!'il~ (penetrations portion)

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Height ofW~=:~: ~~:~: ": ~t(lf iA~~ ~~~'~':~;.:~~~~~for a run '"55*! m 70.F)

Gauge Adjusted Fluid Temperature Pressure for Temperature Kie Kim for P-T Curve P-T Curve (oF)

(ksi*inch 112)

(oF)

(psig) 65.0 76.66 51.10 70 0

65.0 76.66 51.10 70 814 67.0 78.43 52.29 72 834 69.0 80.28 53.52 74 855 71.0 82.20 54.80 76 877 73.0 84.20 56.13 78 900 75.0 86.28 57.52 80 923 77.0 88.44 58.96 82 948 79.0 90.70 60.47 84 973 81.0 93.05 62.03 86 1,000 83.0 95.49 63.66 88 1,028 85.0 98.03 65.35 90 1,056 87.0 100.68 67.12 92 1,086 89.0 103.43 68.95 94 1,118 91.0 106.30 70.86 96 1, 150 93.0 109.28 72.85 98 1, 184 95.0 112.38 74.92 100 1,219 97.0 115.62 77.08 102 1,256 99.0 118.98 79.32 104 1,294

Cooper Nuclear Station PTLR ER 15-019 Revision 1 Page 21 of28 Table 1: CNS Pressure Test (Curve A) P-T Curves for 32 EFPY (continued)

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(ksi*inch 112)

(Of)

(psig) 65.0 84.20 56.13 70 0

65.0 84.20 56.13 70 313 67.0 86.28 57.52 110 313 69.0 88.44 58.96 110 1461 71.0 90.70 60.47 110 1499 73.0 93.05 62.03 110 1539 75.0 95.49 63.66 110 1581 77.0 98.03 65.35 110 1625

Cooper Nuclear Station PTLR ER 15-019 Revision 1 Page 22 of28 Table 2: CNS Core Not Critical (Curve B) P-T Curves for 32 EFPY [7]

Beltline Region P-T Curve P-T Curve Temperature Pressure 70.00 0

70.00 50 70.00 100 70.00 150 70.00 200 70.22 250 81.92 300 84.36 312 84.55 313 91.39 350 99.35 400 106.22 450 114.04 500 123.11 550 130.80 600 137.46 650 143.34 700 148.60 750 153.35 800 157.70 850 161.70 900 165.40 950 168.84 1000 172.07 1050 175.09 1100 177.96 1150 180.65 1200 183.21 1250 185.65 1300

Cooper Nuclear Station PTLR ER 15-019 Revision 1 Page 23 of28 Table 2: CNS Core Not Critical (Curve B) P-T Curves for 32 EFPY (continued)

Plant=

CNS ': <

Component = 'eottoril Head (penetrations portion)

Bottom Head thickness, t = ** * ::6.81'3,,. : inches Bottom Head Radius, R =.

.Hci~ *;;:> inches ART= ;: ' \\.. '*28.. 0 ":> °F======>

. ksi*inch 112 All EFPY Safety Fact~:

1

== *. ** i:}'.~'.~~

Stress Concentration Factor=

3.0<f (bottom head penetrations)

Mm = ' <i.41r '

Temperature Adjustment = ; **.. : 5~0}'.: *<: °F (applied after bolt-up, instrument uncertainty)

Height of Water for a Full Vessel = }:c;*a3{75\\:?~ inches Pressure Adjustment ::: * :.. >*~~Q~§: i\\* psig (hydrostatic pressure head for a full vessel at 70°F)

Pressure Adjustment=

>2~.0 :'. >

.. psig (instrument uncertainty)

Heat Up and Cool Down Rate=

" 1o'Q"'

: <* °F/Hr Gauge Adjusted Fluid Temperature Pressure for Temperature Kie Kim for P-T Curve P-T Curve

("F)

(ksi*inch 112)

("F)

(psig) 65.0 76.66 37.46 70 0

65.0 76.66 37.46 70 582 67.0 78.43 38.35 72 597 69.0 80.28 39.27 74 613 71.0.

82.20 40.23 76 629 73.0 84.20 41.23 78 646 75.0 86.28 42.27 80 664 77.0 88.44 43.36 82 682 79.0 90.70 44.49 84 701 81.0 93.05 45.66 86 721 83.0 95.49 46.88 88 742 85.0 98.03 48.15 90 764 87.0 100.68 49.47 92 786 89.0 103.43 50.85 94 810 91.0 106.30 52.28 96 834 93.0 109.28 53.78 98 859 95.0 112.38 55.33 100 886 97.0 115.62 56.94 102 913 99.0 118.98 58.63 104 942 101.0 122.48 60.38 106 972 103.0 126.12 62.20 108 1,003 105.0 129.92 64.09 110 1,035 107.0 133.86 66.07 112 1,068 109.0 137.97 68.12 114 1,103 111.0 142.25 70.26 116 1, 140 113.0 146.70 72.48 118 1, 178 115.0 151.33 74.80 120 1,217 117.0 156.15 77.21 122 1,258

Cooper Nuclear Station PTLR ER 15-019 Revision 1 Page 24 of28 Table 2: CNS Core Not Critical (Curve B) P-T Curves for 32 EFPY (continued)

Plant=

CNS Component= Upper Vessel ART=

20.0

'F======>

All EFPY Vessel Radius, R =

110.375 inches Nozzle comer thickness, t =

5.753 inches, approximate Ku=

63.45 ksi*inch 112 Kip-applied =

38.90 ksi*inch 112 Crack Depth, a =

1.438 inches Safety Factor=

2.00 Temperature Adjustment =

5.0

°F (applied after bolt-up, instrument uncertainty)

Height of Water for a Full Vessel =

831.75 inches Pressure Adjustment =

30.0 psig (hydrostatic pressure head for a full vessel at 70°F)

Pressure Adjustment =

25.0 psig (instrument uncertainty)

Reference Pressure =

1,000 psig (pressure at which the FEA stress coefficients are valid)

Unit Pressure =

1,563 psig (hydrostatic pressure)

Flange RTNoT =

20.0 OF======>

All EFPY Gauge P*T P*T Fluid Curve Curve Temperature Kie K1p Temperature Pressure (oF)

(ksi*inch 112)

(oF)

(psig) 65.0 84.20 10.37 70 0

65.0 84.20 19.46 70 313 67.0 86.28 20.51 140 313 69.0 88.44 19.26 140 440 71.0 90.70 20.06 140 461 73.0 93.05 20.91 140 482

. 75.0 95.49 21.80 140 505 77.0 98.03 22.73 140 529 79.0 100.68 23.71 140 554 81.0 103.43 24.74 140 581 83.0 106.30 25.82 140 609

. 85.0 109.28 26.95.

140 638 87.0 112.38 28.15 140 668 89.0 115.62 29.40 140 701 91.0 118.98 30.71 140 734 93.0 122.48 32.09 140 770 95.0 126.12 33.54 140 807 97.0 129.92 35.04 140 846 99.0 133.86 36.63 140 887 101.0 137.97 38.30 140 929 103.0 142.25 40.04 140 974 105.0 146.70 41.87 140 1021 107.0 151.33 43.79 140 1071 109.0 156.15 45.80 140 1122 111.0 161.17 47.90 140 1176 113.0 166.39 50.10 140 1233 115.0 171.83 52.39 140 1292

Cooper Nuclear Station PTLR ER 15-019 Revision 1 Page 25 of28 Table 3: CNS Core Critical (Curve C) P-T Curves for 32 EFPY [7]

Curw A Leak Test Temper;~~:t== :>;>i**;~~o<if/\\ °F Curw A Pressure =,J:'.,:~~:.. ~.~~;6\\{;'/i;' psig Unit Pressure=<, <'1;.56.3 ' \\, psig (hydrostatic pressure)

Flange RTNoT = * ;. *2o:o D.:fr °F P-T Curve P-T Curve Temperature Pressure 80.00 0

80.00 50 80.00 100 80.00 150 94.91 200 110.21 250 121.92 300 124.36 312 180.00 313 180.00 350 180.00 400 180.00 450 180.00 500 180.00 550 180.00 600 180.00 650 183.34 7 700 188.60 750 193.35 800 197.70 850 201.70 900 205.40 950 208.84 1000 212.07 1050 215.09 1100 217.96 1150 220.65 1200 223.21 1250 225.65 1300

Plates Welds Nozzles Plates Welds Nozzles Beltline ID Code No.

Lower Shell Plate G-2803-1 Lower Shell Plate G-2803-2 Lower Shell Plate G-2S03-3 Lower Int. Shell Plate G-2802-1 Lower Int. Shell Pl<!te G-2802-2 Lower Int. Shell Plate G-2801-7 Lower Shell Axial Welds 2-233A Lower Shell Axial Welds 2-233B Lower Shell Axial Welds 2-233C Lower Int. Shell Axial Welds l-233A Lower Int. Shell Axial Welds l-233B Lower Int. Shell Axial Welds l-233C Lower/Lower Int. Shell Circ Weld 1-240 Nozzle N-l 6A G-2822 NozzleN-16B G-2822 Beltline ID Code No.

Lower Shell Plate G-2803-1 Lower Shell Plate G-2803-2 Lower Shell Plate G-2803-3 Lower Int. Shell Plate G-2802-1 Lower Int. Shell Plate G-2802-2 Lower Int. Shell Plate G-2801-7 Lower Shell Axial Welds 2-233A Lower Shell Axial Welds 2-233B Lower Shell Axial Welds 2-233C Lower Int. Shell Axial Welds l-233A Lower Int. Shell Axial Welds l-233B Lower Int. Shell Axial Welds l-233C Lower/Lower Int. Shell Circ Weld 1-240 Nozzle N-16A G-2822 Nozzle N-168 G-2822 Table 4: CNS ART Calculations for 32 EFPY [6]

Heat No.

Flux Type Initial Cu Ni CF RT NOT

(°F)

(wt%)

(°F)

C2274-l 14.0 0.20 0.68 153.0 C2307-l 0.0 0.21 0.73 162.8 C2274-2

-8.0 0.20 0.68 153.0 C2331-2 10.0

((.]

n*i n*i C2307-2

-20.0 n*i n*n n*i C2407-I

-10.0 0.13 0.65 92.3 12420 LINDE 1092

-50.0 0.270 l.035 254.4 12420 LINDE 1092

-50.0 0.270 1.035 254.4 12420 LINDE 1092

-50.0 0.270 1.035 254.4 27204/12008 LINDE 1092

-50.0 0.219 0.996 231.1 27204/12008 LINDE 1092

-50.0 0.219 0.996 231.1 21935 LINDE 1092

-50.0 0.183 0.704 172.2 EV-26067

-10.0 0.13 0.65 92.3 EV-26067 10.0 0.16 0.62 118.5 Fluence Data Heat No.

Wall Thickness (in.)

Fluence Attenuation Fluence at ID at l/4t Full l/4t (n/cm2) e--0.24x (n/cm2)

C2274-1 6.375 1.59 l.09E+I8 0.68 7.44£+17 C2307-l 6.375 1.59 I.09E+I8 0.68 7.44E+l7 C2274-2 6.375 1.59 I.09E+I8 0.68 7.44E+I7 C2331-2 5.375 1.34 1.4IE+!8 0.72

!.02E+l8 C2307-2 5.375 1.34 1.41E+l8 0.72 1.02E+l8 C2407-l 5.375 1.34 J.41E+l8 0.72 l.02E+18 12420 6.375 1.59 I.07E+l8 0.68 7.30E+l7 12420 6.375 1.59

!.07E+l8 0.68 7.30E+l7 12420 6.375 1.59

!.07E+l8 0.68 7.30E+17 27204/12008 5.375 1.34 8.11E+l7

0.72 5.87E+l7 27204/12008 5.375 1.34 8.llE+l 7 0.72 5.87E+l7 27204/12008 5.375 1.34 8.11E+l7 0.72 5.87E+l7 21935 5.375 1.34 I.09E+l8 0.72 7.90E+l7 EV-26067 5.375 1.34 2.94E+!7 0.72 2.13E+l7 EV-26067 5.375 1.34 2.94E+l7 0.72 2.13£+17 EPRI Proprietary Information Cooper Nuclear Station PTLR ER 15-019 Revision 1 Page 26 of28 ARTNDT Margin Terms Total Mare: in

(°F)

O"A (°F)

O"; (*F)

(°F) 55.l 17.0 0.0 34.0 58.6 17.0 0.0 34.0 55.l 17.0 0.0 34.0 63.0 8.5 0.0 17.0 108.8 8.5 0.0 17.0 38.8 17.0 0.0 34.0 90.8 28.0 0.0 56.0 90.8.

28.0 0.0 56.0 90.8 28.0 0.0 56.0 73.7 28.0 0.0 56.0 73.7 28.0 0.0 56.0 73.7 28.0 0.0 56.0 63.9 28.0 0.0 56.0 16.5 8.3 0.0 16.5 21.2 10.6 0.0 21.2 Fluence Factor, FF j<0.?8-0.IOlogQ 0.360 0.360 0.360 0.421 0.421 0.421 0.357 0.357 0.357 0.319 0.319 0.319 0.371 0.179 0.179 (such information is marked with double braces "((]}"and a bar in the right-hand margin)

ART

(°F) 103.1 92.6 81.1 90.0 105.8 62.8 96.8 96.8 96.8 79.7 79.7 79.7 69.9 23.0 52.4

Appendix A Cooper Nuclear Station PTLR ER 15-019 Revision I Page 27 of28 COOPER REACTOR VESSEL MATERIALS SURVEILLANCE PROGRAM In accordance with 10 CFR 50, Appendix H, Reactor Vessel Material Surveillance Program Requirements [18], two surveillance capsules were removed from the CNS reactor vessel in 1985 at 6.8 EFPY and 1991at11.2 EFPY [20, Attachment 3]. The surveillance capsules contained flux wires for neutron fluence measurement, Charpy V-Notch impact test specimens and uniaxial tensile test specimens fabricated using materials from the vessel materials within the core beltline region.

CNS is currently committed to use the BWRVIP ISP, and has made a licensing commitment to use the ISP for CNS during the period of extended operation. The BWRVIP ISP meets the requirements of 10 CFR 50, Appendix H, for Integrated Surveillance Programs, and has been approved by the NRC.

Nebraska Public Power District committed to use the ISP in place of its existing surveillance programs in the amendments issued by the NRC regarding the implementation of the Boiling Water Reactor Vessel and Internals Project Reactor Pressure Vessel Integrated Surveillance Program, dated October 31, 2003 (23]. Under the ISP, a capsule was scheduled for removal in 2003 but removal has been deferred to approximately 2017 at 32 EFPY (21]. CNS recently transitioned to 24 month refueling cycles during "even" years so the next capsule removal will occur in 2018 to align with a plant refueling outage as allowed by the ISP [21]. Additionally, CNS served as a host plant for three of the nine surveillance capsules irradiated as part of the Supplemental Surveillance Program; the SSP-A, SSP-B, and SSP-C capsules were removed from CNS and tested in 2003 [21] The surveillance capsules contained flux wires for neutron fluence measurement, Charpy V-Notch impact test specimens and uniaxial tensile test specimens fabricated using materials from the vessel materials within the core beltline region. CNS continues to be a host plant under the ISP. One additional standby Cooper capsule is currently scheduled to be removed and tested under the ISP during the license renewal period in approximately 2029 at 40 EFPY [21].